Magnesium transport extraction of transuranium elements from LWR fuel

ABSTRACT

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl 2  and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800° C. to about 850° C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl 2  having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO 2 . The Ca metal and CaCl 2  is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

CONTRACTUAL ORIGIN OF THE INVENTION

The United States Government has rights in this invention pursuant toContract No. W-31-109-ENG-38 between the U.S. Department of Energy andThe University of Chicago representing Argonne National Laboratory.

BACKGROUND OF THE INVENTION

This invention relates to a pyrochemical process for converting spentoxide nuclear fuel from a light water reactor to metal and forseparating plutonium and higher actinide metals such as americium,neptium and curium from the bulk uranium. Because the end product is foruse in a integrated fast reactor (IFR), high decontamination of theseparate streams from fission products is not a prime concern nor is thetotal separation of plutonium, americium, neptunium and curium(hereinafter transuranic elements) from the bulk uranium. Thetransuranic or transuranium elements will be used to make core fuel fora liquid metal fast breeder (LMFBR) particularly of the new IFR orintegrated fast reactor type. Because of the purpose for which thisreprocessed fuel will be used, some uranium can accompany the plutoniumstream since the uranium to plutonium ratio in a LMFBR fuel is in therange of 2-3.5:1. The bulk uranium or uranium rich product stream is tobe stored for later use, for instance as a uranium source for breederblankets in a liquid metal fast breeder reactor (LMFBR), when and ifsuch fast reactors are commercially viable. A goal of the process is toremove approximately 90% of the transuranic or transuranium actinidesfrom the uranium so that the transuranic actinides can be used as corefuel and the remaining uranium can be used as blanket material.

Accordingly, it is an object of the invention to provide a process forseparating transuranic or transuranium actinide values from spent oxidenuclear fuel while reducing the amount of nuclear waste material whichhas to be treated and stored.

Another object of the invention is to provide a process using variouscombinations of alloys, salts and liquid magnesium selectively toseparate uranium from the transuranic values present in spent nuclearoxide fuel and to reuse the salts and the magnesium many times in orderefficiently to separate the desired values while producing a very smallamount of nuclear waste.

A still further object of the invention is to provide a process ofseparating transuranium actinide values from uranium values present inspent nuclear oxide fuels containing rare earth and noble metal fissionproducts as well as other fission products, comprising reducing theoxide fuel with Ca metal in the presence of Ca halide and a U-Fe alloywhich is liquid at about 800° C. to dissolve uranium metal and the noblemetal fission products and transuranium actinide metals and rare earthfission product metals leaving Ca halide having CaO and fission productsof alkali metals and the alkali earth metals and iodine dissolvedtherein, separating the Ca halide and CaO and the fission productscontained therein from the U-Fe alloy and the metal values dissolvedtherein, contacting the U-Fe alloy having dissolved therein reducedmetals from the spent nuclear fuel with Mg metal to transfertransuranium actinide metals and rare earth metals to the liquid Mgmetal leaving the uranium and noble metal fission products in the U-Fealloy, thereafter separating the Mg and the metals dissolved thereinfrom the U-Fe alloy and the metals dissolved therein, and distilling theMg from the transuranium actinide and rare earth metals.

The invention consists of certain novel features and a combination ofparts hereinafter fully described, illustrated in the accompanyingdrawings, and particularly pointed out in the appended claims, it beingunderstood that various changes in the details may be made withoutdeparting from the spirit, or sacrificing any of the advantages of thepresent invention.

BRIEF DESCRIPTION OF THE DRAWING

For the purpose of facilitating an understanding of the invention, thereis illustrated in the accompanying drawings a preferred embodimentthereof, from an inspection of which, when considered in connection withthe following description, the invention, its construction andoperation, and many of its advantages should be readily understood andappreciated.

FIG. 1 is a schematic diagram for illustrating the process of thepresent invention.

DETAILED DESCRIPTION OF THE INVENTION

The inventive process begins with spent nuclear oxide fuel from a lightwater reactor which has been mechanically declad such that the productof the decladding is the oxide pellet and/or oxide particulate which isused as a process feed. The process feed is introduced into an oxidereduction vessel 10 and particularly into a crucible in the vessel 10 bymeans of an inlet line 12. In the vessel 10 is a two-phase systemconsisting of a uranium-iron liquid alloy introduced through inlet line13 on top of which are liquid calcium metal introduced through inletline 14 and a liquid calcium chloride salt introduced through inlet line15, the entire vessel 10 being maintained at a temperature in the rangeof from about 750° C. to about 850° C. with the preferred temperaturebeing about 850° C.

Any calcium halide salt or mixture is adequate, providing the meltingpoint is less than about 750° C.; however, CaCl₂ should be present as amajority constituent and the preferred salt is 85% by weight CaCl₂ withthe balance CaF₂. As the oxide fuel is introduced into the vessel 10 andmixed by mechanism (not shown), the oxides are reduced by reaction withthe liquid calcium metal to form calcium oxide which dissolves in thecalcium chloride salt, producing the uranium and transuranic metalsalong with noble metal fission products. Although uranium has a meltingpoint of about 1150° C. when present in an alloy with Fe, wherein theuranium concentration is in the range of from about 84% by weight toabout 96% by weight, the alloy is liquid at 850° C. and where uranium ispresent in the range of from about 87% by weight to about 94% by weight,the alloy is liquid at 800° C. Alkali metal and alkaline earth metal andiodine fission products which are reduced by the calcium dissolve aschlorides in the calcium chloride salt phase whereas the transuranicactinides including plutonium dissolve in the U-Fe alloy phase alongwith rare earth fission product metals and the noble metal fissionproducts.

The preferred alloy used in this phase of the separation is a uranium88% by weight-Fe alloy to which is added uranium from the reduced fueluntil uranium is present in an amount of about 93 to 94% by weight. Goodreductions of the oxide fuel take place above 750° C., but the higherthe temperature, the greater the vapor pressure and more corrosive thereactants. Preferably, the reduction takes place about 850° C. Afterreduction of the oxide fuel, a three phase system exists within thevessel 10. Uranium-iron alloy and the noble metal fission products andthe transuranic actinides along with the rare earth metals are dissolvedin the alloy. The salt phase includes the dissolved calcium oxide whichis the product of the reduction of the oxide fuel along with the alkalimetal and alkaline earth metal and iodine fission products which migrateto the salt as chlorides. After the reduction in vessel 10 is complete,the calcium chloride salt containing dissolved calcium oxide istransported by means (not shown) to a calcium regenerator vessel 20.

The oxide fuel reduction vessel 10 is provided with a top so as to closethe vessel 10 during the reduction. The calcium regenerator vessel 20includes an electrochemical mechanism (not shown) having a carbonelectrode 22 connected to a source of electrical power (not shown). Aline 17 from the oxide fuel reduction vessel 10 leads to the vessel 20wherein it receives the calcium chloride salt having the calcium oxidedissolved therein along with the alkali metal and alkaline earth fissionproduct chlorides which have dissolved in the calcium salt during thereduction of the oxide fuel in vessel 10. In order to be certain thatall of the calcium is transferred from the vessel 10, a small portion ofthe uranium-iron actinide-containing alloy is also transferred to thecalcium regenerator 20. A calcium-zinc alloy could be used as analternative to the uranium-iron alloy in order to accumulate the calciummetal produced during operation of the electrochemical mechanism in thevessel 20. The mechanism includes a liquid metal, preferably Zn, cathodeand a porous screen surrounding the carbon anode to prevent particulatecarbon particles from contaminating the CaCl₂ salt.

Upon operation of the electrochemical mechanism in a well known manner,carbon monoxide and carbon dioxide are produced and exit by line 22 asthe carbon electrode is consumed while calcium metal is produced duringthe reduction of calcium oxide at the electrode and recycled to line 14while the salt is recycled to line 15. The electrochemical method hereinuses a molten Zn cathode and a porous screen surrounding the carbonanode to prevent carbon particulates from contaminating the CaCl₂-containing salt. The calcium metal produced by the electrolyticalprocess is taken up either in the uranium-iron alloy from the vessel 10or by the alternative calcium-zinc alloy, into either of which thecalcium will dissolve. If a calcium-zinc alloy is used, then the zincmust be retorted from the alloy involving another step. Accordingly, thepreferred alloy used to accumulate the calcium metal produced during thedestruction of the carbon electrode is the uranium-iron alloy used inthe reduction vessel 10.

The liquid metal alloy phase left in the vessel 10 after the calciumchloride salt has been pumped into the calcium regenerator 20 consistsof the original U-Fe alloy with additional uranium reduced from theoxide fuel in addition to the transuranium actinides, the noble metalsfission products, the rare earth fission products, with only the alkalimetal, alkaline earth metals and iodine fission products transferring tothe CaCl₂ -CaF₂ salt. The metal alloy from the reduction vessel 10 exitsthe vessel via an exit line 18 and enters a magnesium extraction vessel30. The magnesium extraction vessel 30 has a magnesium input line 31 anda magnesium exit line 32. A product line 33 conducts the metal alloyproduct, as will be explained, to storage or further processing. Themagnesium extraction vessel 30 is operated at a temperature of about800° C. and in it magnesium from the inlet line 31 is intimately mixedwith the uranium-iron metal alloy containing, as heretofore stated, thetransuranium actinides, the noble metal fission products, the rare earthfission products, all of which are liquid and dissolved in the metaluranium-iron alloy. At this point the uranium content in theuranium-iron alloy is greater than the uranium content in the alloy invessel 10 prior to the oxide reduction but, it is critical that theuranium does not exceed about 94% by weight of the alloy in order forthe alloy to remain liquid at 800° C., the preferred magnesiumextraction temperature. In the event that the uranium content increasesto about 96% by weight, then the magnesium extraction must take place ata higher temperature, such as 850° C., or additional iron must be addedin order to reduce the weight percent of uranium in the alloy. In anyevent, after mixing, the magnesium takes up the plutonium and otherminor actinides along with the rare earth fission product and isseparated and pumped, by means not shown, through line 32 to a magnesiumdistillation apparatus 40. Remaining in the vessel 30 is a uranium-ironalloy containing the noble metal fission products which is generallystored for later use in as blanket material in a breeder reactor.

The magnesium containing the plutonium, rare earth fission products andother actinides is distilled in the apparatus 40 with the magnesiumbeing recycled via line 3 to the magnesium extraction vessel 30, and theplutonium, other actinides and rare earth fission products are pumpedthrough line 41 to an electrorefiner for further processing as corematerial for an IFR. The magnesium distillation preferably takes placeat about 950° C. It is important to understand that when the metal alloyleaves the vessel 10 some uranium-iron alloy remains in the vessel for anew batch of oxide fuel. Fresh iron can be added to return theuranium-iron content to the original concentration of about 88% byweight uranium. It should be understood that the oxide reduction makesuse of the eutectic valley which exists for uranium-iron alloys whereinthe uranium is present at approximately 90% by weight and the iron ispresent at about 10% by weight so as to be able to operate the processin the 800° C. to 850° C. range. In this range, the uranium content canvary from 84 % by weight to about 96% by weight. As before stated, thereduction reaction takes place better at higher temperatures, with 750°C. being the lowest temperature at which the oxide reduction takesplace. On the other hand, with increased temperatures, the vaporpressures of the materials increase and the corrosion rates of thematerials increase thereby presenting handling problems so that atrade-off exists between the higher temperatures which are good for thereduction reactions and the lower temperatures needed for easierhandling.

An important aspect of the invention is that no additional radioactivewaste is created by the separation of the transuranic actinide values asdescribed herein. Because both the calcium salt and the magnesium arereused in this process, little radioactive waste is created by thereprocessing of the exhausted fuel. The uranium-iron noble metal alloyis stored for later use as blanket material and therefore long termstorage is not required. Because the magnesium is continually distilled,and reused for extraction, it is a continuous cycle, producing little,if any, radioactive waste.

The CaCl₂ -containing salt may be used in up to fifty batches beforesufficient quantities of alkali earth metal and alkali metal fissionproducts have accumulated in the salt such that the heat generatedexceeds present regulatory limitations for storage of this radioactivematerial. Accordingly, a significant number of batches of oxide fuel canbe processed by this process without using additional calcium chloridesalt or contributing to the amount of nuclear waste material which mustbe safely stored.

As is understood from the explanation herein, the process is essentiallya batch process which may be repeated a number of times. The size ofeach batch cycle is limited to the amount of plutonium which may beconcentrated in pure form from each batch. The reduction batch may belarger but the salt transport portion is limited to approximately 3kilograms of plutonium in any single batch because of criticalityconsiderations. Because the amount of plutonium produced in the oxidefuel, which is about 0.9 weight percent and the 3 kilogram limitation,each batch of oxide fuel from a LWR reactor reprocessed by the inventiveprocess in about 333 kilograms.

While there has been disclosed what is considered to be the preferredembodiment of the present invention, it is understood that variouschanges in the details may be made without departing from the spirit, orsacrificing any of the advantages of the present invention.

The embodiments of the invention in which an exclusive property orprivilege is claimed are defined as follows:
 1. A process of separatingtransuranium actinide values from uranium values present in spentnuclear oxide fuels containing rare earth and noble metal fissionproducts as well as fission products of alkali metals, the alkalineearth metals and iodine, comprising reducing the oxide fuel with Cametal in the presence of Ca halide and a U-Fe alloy which is liquid atabout 800° C. to dissolve uranium values and the noble metal fissionproducts and transuranium actinide values and rare earth fissionproducts leaving Ca halide having CaO and fission products of alkalimetals and the alkaline earth metals and iodine dissolved therein,separating the Ca halide and CaO and the fission products containedtherein from the U-Fe alloy and the metal values dissolved therein,contacting the U-Fe alloy having dissolved therein reduced metals fromthe spent nuclear fuel with liquid Mg metal to transfer transuraniumactinide values and rare earth metals to the liquid Mg metal leaving theuranium and noble metal fission products in the U-Fe alloy, thereafterseparating the Mg and the metal dissolved therein from the U-Fe alloyand the metals dissolved therein, and distilling the Mg from thetransuranium actinide values and rare earth metals.
 2. The process ofclaim 1, wherein the Ca halide includes CaCl₂.
 3. The process of claim1, wherein the Ca halide is a combination of CaCl₂ and CaF₂ having amelting point less than 750° C.
 4. The process of claim 3, wherein thecombination of Ca halides is about 85% by weight CaCl₂ and the balanceCaF₂.
 5. The process of claim 1, wherein uranium is present in theuranium-Fe alloy in the range of from about 84% by weight to about 96%by weight.
 6. The process of claim 1, wherein the uranium is present inthe range of from about 87% by weight to about 94% by weight.
 7. Theprocess of claim 1, wherein the temperature is maintained at not lessthan about 750° C.
 8. The process of claim 7, wherein the temperatureduring oxide reduction and during contact with the liquid Mg ismaintained in the range of from about 750° C. to about 850° C.
 9. Theprocess of claim 7, wherein the reduction of spent nuclear oxide fuelwith Ca metal takes place at about 850° .
 10. The process of claim 9,wherein the extraction of transuranium actinide values and rare earthmetals from the U-Fe alloy to Mg takes place at a temperature of about800° C.
 11. A process of separating transuranium actinide values fromuranium values present in spent nuclear oxide fuels containing rareearth and noble metal fission products as well as fission products ofalkali metals, the alkaline earth metals an iodine, comprising reducingthe oxide fuel with Ca metal in the presence of Ca halide and a U-Fealloy which is liquid at about 800° C. to dissolve uranium values andthe noble metal fission products and transuranium actinide values andrare earth fission product metals leaving Ca halide having CaO andfission products of alkali metals and the alkaline earth metals andiodine dissolved therein, separating the Ca halide with the CaO and thefission products contained therein from the U-Fe alloy and the metalvalues dissolved therein and electrolytically contacting the calciumsalts with a carbon electrode to reduce the CaO and Ca metal whileconverting the carbon electrode to CO and CO₂ and thereafter recyclingthe Ca metal and Ca halide salt to reduce additional oxide fuel,contacting the liquid U-Fe alloy having dissolved therein reduced metalsfrom the spent nuclear fuel with liquid Mg metal to transfertransuranium actinide values and rare earth metals to the liquid Mgmetal leaving the uranium and noble metal fission products in the U-Fealloy, thereafter separating the liquid Mg and the metals dissolvedtherein from the U-Fe alloy and the metals dissolved therein, distillingthe Mg from the transuranium actinide and rare earth metals, andrecontacting the U-Fe alloy with liquid Mg metal a sufficient number oftimes until not less than about 99% by weight of the transuraniumactinide values have been removed from the U-Fe alloy.
 12. The processof claim 11, wherein the Ca halide is at least 50% CaCl₂.
 13. Theprocess of claim 12, wherein the Ca halide is 85% by weight CaCl₂ andthe balance CaF₂.
 14. The process of claim 11, wherein uranium ispresent in the uranium-Fe alloy in the range of from about 84% by weightto about 96% by weight.
 15. The process of claim 11, wherein thereduction of spent nuclear oxide fuel with Ca metal takes place at about850° C., and the extraction of transuranium metals and rare earth metalsfrom the U-Fe alloy to Mg takes place at a temperature at about 800° C.16. A process of separating transuranium actinide values from uraniumvalues present in spent nuclear oxide fuels containing rare earth andnoble metal fission products as well as fission products of alkalimetals, alkaline earth metals and iodine, comprising reducing the oxidefuel with Ca metal in the presence of Ca halide containing predominantlyCaCl₂ and a U-Fe alloy having not less than about 84% by weight uraniumwhich is liquid at about 800° C. to dissolve uranium values and thenoble metal fission products and transuranium actinide values and rareearth fission products leaving Ca halide having CaO and fission productsof alkali metals and the alkaline earth metals and iodine dissolvedtherein, separating the Ca halide with the CaO and the fission productscontained therein from the U-Fe alloy and the metal values dissolvedtherein and electrolytically contacting the calcium salts with a carbonelectrode to reduce the CaO to Ca metal while converting the carbonelectrode to CO and CO2 and thereafter recycling the Ca metal and Cahalide salt to reduce successive batches of spent nuclear oxide fuel,contacting the liquid U-Fe alloy having dissolved therein reduced metalsfrom the spent nuclear fuel with liquid Mg metal to transfertransuranium actinide values and rare earth metals to the liquid Mgmetal leaving the uranium and noble metal fission products in the U-Fealloy, thereafter separating the liquid Mg and the metals dissolvedtherein from the U-Fe alloy and the metals dissolved therein, distillingthe Mg from the transuranium actinide and rare earth metals,recontacting the U-Fe alloy with liquid Mg metal a sufficient number oftimes until not less than about 99% by weight of the transuraniumactinide values have been removed from the U-Fe alloy, and thereaftercontacting a portion of the U-Fe alloy with another batch of spentnuclear oxide fuel in the presence of Ca metal and Ca halide salt at atemperature not less than about 800° C.
 17. The process of claim 16,wherein the Ca halide salt has a melting point of about 650° F.
 18. Theprocess of claim 16, wherein the Ca halide salt is about 85% by weightCaCl₂.
 19. The process of claim 16, wherein the oxide reduction takesplace at about 850° C. and contact with the liquid Mg takes place atabout 800° C.
 20. The process of claim 16, wherein the U in the U-Fealloys is present in the range of from about 84% by weight to about 96%by weight.